UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D.C. 20555
July 2, 1979
IE Bulletin No. 79-14
SEISMIC ANALYSES FOR AS-BUILT SAFETY-RELATED PIPING SYSTEMS
Description of Circumstances:
Recently two issues were identified which can cause seismic analysis of
safety-related piping systems to yield nonconservative results. One issue
involved algebraic summation of loads in some seismic analyses. This was
addressed in show cause orders for Beaver Valley, Fitzpatrick, Maine Yankee
and Surry. It was also addressed in IE Bulletin 79-07 which was sent to all
power reactor licensees.
The other issue involves the accuracy of the information input for seismic
analyses. In this regard, several potentially unconservative factors were
discovered and subsequently addressed in IE Bulletin 79-02 (pipe supports)
and 79-04 (valve weights). During resolution of these concerns, inspection
by IE and by licensees of the as-built configuration of several piping
systems revealed a number of nonconformances to design documents which could
potentially affect the validity of seismic analyses. Nonconformances are
identified in Appendix A to this bulletin. Because apparently significant
non- conformances to design documents have occurred in a number of plants,
this issue is generic.
The staff has determined, where design specifications and drawings are used
to obtain input information for seismic analysis of safety-related piping
systems, that it is essential for these documents to reflect as-built
configurations. Where subsequent use, damage or modifications affect the
condition or configuration of safety-related piping systems as described in
documents from which seismic analysis input information was obtained, the
licensee must consider the need to re-evaluate the seismic analyses to
consider the as-built configuration.
.
IE Bulletin No. 79-14 July 18, 1979
Revision 1 Page 2 of 3
Action to be taken by Licensees and Permit Holders:
All power reactor facility licensees and construction permit holders are
requested to verify, unless verified to an equivalent degree within the last
12 months, that the seismic analysis applies to the actual configuration of
safety-related piping systems. The safety related piping includes Seismic
Category I systems as defined by Regulatory Guide 1.29, "Seismic Design
Classification" Revision 1, dated August 1, 1973 or as defined in the
applicable FSAR. The action items that follow apply to all safety related
piping 2 1/2-inches in diameter and greater and to seismic Category I
piping, regardless of size which was dynamically analyzed by computer. For
older plants, where Seismic Category I requirements did not exist at the
time of licensing, it must be shown that the actual configuration of
safety-related systems, utilizing 2 1/2 inches in diameter and greater,
meets design requirements.
Specifically, each licensee is requested to:
1. Identify inspection elements to be used in verifying that the seismic
analysis input information conforms to the actual configuration of
safety-related systems. For each safety-related system, submit a list
of design documents, including title, identification number, revision,
and date, which were sources of input information for the seismic
analyses. Also submit a description of the seismic analysis input
information which is contained in each document. Identify systems or
portions of systems which are planned to be inspected during each
sequential inspection identified in Items 2 and 3. Submit all of this
information within 30 days of the date of this bulletin.
2. For portions of systems which are normally accessible*, inspect one
system in each set of redundant systems and all nonredundant systems
for conformance to the seismic analysis input information set forth in
design documents. Include in the inspection: pipe run geometry; support
and restraint design, locations, function and clearance (including
floor and wall penetration). embedments (excluding those covered in IE
Bulletin 79-02); pipe attachments; and valve and valve operator
locations and weights (excluding those covered in IE Bulletin 79-04).
Within 60 days of the date of this bulletin, submit a description of
the results of this inspection. Where nonconformances are found which
affect operability of any system, the licensee will expedite completion
of the inspection described in Item 3.
* Normally accessible refers to those areas of the plant which can be
entered during reactor operation.
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IE Bulletin No. 79-14 July 2, 1979
Page 3 of 3
3. In accordance with Item 2, inspect all other normally accessible
safety-related systems and all normally inaccessible safety-related
systems. Within 120 days of the date of this bulletin, submit a
description of the results of this inspection.
4. If nonconformances are identified:
A. Evaluate the effect of the nonconformance upon system operability
under specified earthquake loadings and comply with applicable
action statements in your technical specifications including
prompt reporting.
B. Submit an evaluation of identified nonconformances on the validity
of piping and support analyses as described in the Final Safety
Analysis Report (FSAR) or other NRC approved documents. Where you
determine that reanalysis is necessary, submit your schedule for:
(i) completing the reanalysis, (ii) comparisons of the results to
FSAR or other NRC approved acceptance criteria and (iii)
submitting descriptions of the results of reanalysis.
C. In lieu of B, submit a schedule for correcting nonconforming
systems so that they conform to the design documents. Also submit
a description of the work required to establish conformance.
D. Revise documents to reflect the as-built conditions in plant, and
describe measures which are in effect which provide assurance that
future modifications of piping systems, including their supports,
will be reflected in a timely manner in design documents and the
seismic analysis.
Facilities holding a construction permit shall inspect safety-related
systems in accordance with Items 2 and 3 and report the results within 120
days.
Reports shall be submitted to the Regional Director with copies to the
Director of the Office of Inspection and Enforcement and the Director of the
Division of Operating Reactors, Office of Nuclear Reactor Regulation,
Washington, D.C. 20555.
Approved by GAO (R0072); clearance expires 7/31/80. Approval was given under
a blanket clearance specifically for generic problems.
. APPENDIX A
PLANTS WITH SIGNIFICANT DIFFERENCES BETWEEN ORIGINAL DESIGN AND AS-BUILT
CONDITION OF PIPING SYSTEMS
Plant Difference Remarks
Surry 1 Mislocated supports. As built condition
Wrong Support Type. caused majority of pipe
Different Pipe Run overstress problems, not
Geometry. algebraic summation.
Beaver Valley Not specifically identified. As built condition
Licensee reported "as-built resulted in both pipe and
conditions differ signifi- support overstress.
cantly from original design."
Fitzpatrick IE inspection identified Licensee is using as
differences similar to built configuration
Surry. for reanalysis.
Pilgrim Snubber sizing wrong. Plant shutdown to restore
Snubber pipe attachment original design condition.
welds and snubber support
assembly nonconformances.
Brunswick 1 and 2 Pipe supports undersize. Both units shutdown to
restore original design
condition.
Ginna Pipe supports not built Supports were repaired
to original design. during refueling outage.
St. Lucie Missing seismic supports. Install corrected
Supports on wrong piping. supports before start
up from refueling.
.
Page 2 APPENDIX A
Plant Difference Remarks
Nine Mile Point Missing seismic supports. Installed supports before
startup from refueling.
Indian Point 3 Support location and Licensee performing as
support construction built verification to be
deviations. completed by July 1.
Davis-Besse Gussets missing from main Supports would be over-
Steam Line Supports. stressed. Repairs will be
completed prior to start-
up.